Method of producing zirconium base alloys

ABSTRACT

A PROCESS FOR PRODUCING A ZIRCONIUM BASE ALLOY MEMBER HAVING IMPROVED TENSILE STRENGTH AND DUCTILITY PROPERTIES PRIOR TO NUCLEAR RADIATION BY FIRST COLD WORKING THE MEMBER TO INDUCE A 7% TO 20% STRAIN. THE MEMBER IS THEN HEAT TREATED AT TEMPERATURES LOWER THAN THE USUAL ANNEALING TEMPERATURES.

United States Patent 3,567,522 METHOD OF PRODUCING ZIRCONIUM BASE ALLOYS Kenneth C. Thomas and Robert J. Allio, Pittsburgh, Pa.,

assignors to Westinghouse Electric Corporation, Pittsburgh, Pa. No Drawing. Filed Dec. 15, 1965, Ser. No. 514,129 Int. Cl. C22f 1/18 US. Cl. 14811.5 3 Claims ABSTRACT OF THE DISCLOSURE A process for producing a zirconium base alloy member having improved tensile strength and ductility properties prior to nuclear radiation by first cold working the member to induce a 7% to 20% strain. The member is then heat treated at temperatures lower than the usual annealing temperatures.

This invention relates to zirconium alloys having improved tensile properties and ductility.

Of the various materials used in a nuclear reactor, zirconium base alloys are employed primarily because of the low neutron absorption of the element zirconium. Typical zirconium base alloys are known as zircaloy-Z and zircaloy-4. The nomial composition of zircaloy-2 includes 1.5 Weight percent tin, 0.12% iron, 0.10% chromium, 0.05% nickel, and the balance zirconium with incidental impurities. Zirconium-4 has a nominal composition of 1.5 weight percent tin, 0.2% iron, 0.1% chromium, 0.007% nickel, and the balance zirconium with incidental impurities. In the nuclear industry these materials are used as fuel element cladding, structural parts, and supports, because of their low neutron absorption, good mechanical strength, heat resistance, and excellent corrosion resistance at reactor operating temperatures.

In a nuclear environment, however, the performance of these alloys may be severely diminished by the lattice damage induced by radiation, in particular, the elongation may drop to about 1%. This damage may result in an increase in the rate of precipitation of other phases thereby giving rise to the loss of ductility. Generally, the loss of ductility increases with the increase in radiation exposure, resulting in very high susceptibility to failure in a brittle manner after extended service time. Consequently, in fuel elements clad with zirconium alloys, a high probability of metal failure occurs due to the buildup of fission gases within the fuel tubes.

Associated with the foregoing is the fact that exposure to radiation damages the lattice when neutrons collide with and displace the atoms of the zirconium base alloys. That is particularly true at lower temperatures of reactor operation, such as below 750 F., because significant amounts of damage are retained and in fact tend to increase with extended exposure times to radiation.

Further, zirconium base alloys have a dual brittleness problem which includes not only the lattice damage due to collision of neutrons and other fission products, but also the problem of hydrogen embrittlement. When the alloy member absorbs hydrogen beyond its hydrogen solubility limit, zirconium hydride (ZrH) may be precipitated in an undesirable pattern. The resulting loss of tensile strength and ductility makes the fuel tube cladding particularly susceptible to the pressures of fission gas build-up and the precipitated zirconium hydrides may serve as initiation sites for cracks on the outer surface of the cladding.

It has been found that the foregoing problems may be alleviated by providing the alloy tube members with the highest possible ductility before they are placed into re- 3,567,522 Patented Mar. 2, 1971 actor service. However, the highest possible ductility which is obtained by fully annealing the alloy is not the complete answer to the problem because the necessary specifications require that the alloy must have higher values of elongation and yield strength than are ordinarily obtained by fully annealing the alloy members. In metallurgical practice it is the usual phenomenon that for a given alloy composition at about its optimum condition, any improvement in ductility is accompanied by a loss in yield strength or an increase in yield strength is obtained at the expense of ductility. By the present invention the desired combination of higher yield strength together with the greatly improved ductility is obtained unexpectedly by preliminarily stressing the alloy material by imparting a 7% to 20% tensile strain and then heat treating the material for predetermined periods of time at heat treating temperatures lower than the usual annealing temperatures.

Accordingly, it is a general object of this invention to provide a method for producing zirconium alloys having increased yield strengths and improved ductility in terms of higher elongation values.

It is another object of this invention to provide a method for producing zirconium alloys which method includes a combination of cold working and annealing for obtaining a required high yield strength and at the same time providing a considerably higher ductility.

Finally, it is an object of this invention to overcome the foregoing problems and desiderata in a simple and effective manner.

In accordance with the present invention it has been discovered that alloys of zirconium, such as zircaloy-2 and zircaloy-4, may be imbued with improved ductility and higher tensile properties by first cold working the alloy member to induce a 7% to 20% strain in the alloy material at room temperature and up to about 600 C., and then annealing the member at a temperature ranging from 250 to 600 C. for a time period varying from a minimum of several hours to about one minute, the time and temperature being correlated so that the longer period is required for the lowest temperatures. The optimum results are derived when the strain is applied at from room temperature to 300 C., and the annealing is effected for 15 to 45 minutes at 250 C. to 350 C.

An alloy of the present invention may be readily prepared by melting a relatively pure grade of zirconium, tin, iron, chromium, and nickel in weighed increments in an induction furnace or in an arc melting furnace, which for example, may employ a water cooled copper crucible and a tungsten-tipped electrode. The charge is melted under an atmosphere of purified argon or helium to prevent contamination of the melt. The resulting melt is solidified into an ingot, which may be remelted in accordance with the process set forth in US. Pat. No. 3,072,982. The resulting ingot is heated to 1800 F, and forged into a slab which may then be hot rolled at a temperature of 1550 F. into a sheet rod or strip that may be fabricated or drawn into tubes as desired.

A typical sample of an annealed hot rolled strip having the analysis of zirca1oy-2 or zircaloy-4 as listed above may have a yield strength of 40,000 to 45,000 p.s.i. and only moderate ductility which is insufiicient strengh for many applications in a nuclear environment.

By a combination of cold working and annealing it has been possible to obtain zirconium base alloy material with higher yield strengths as well as with a considerably higher ductility.

The zirconium base alloys of the present invention have proven to be exceptionally satisfactory when strained or worked from 7% to 20% at temperatures ranging from about 25 to 600 C. followed by an anice ealing heat treatment at relatively low temperatures ranging from 250 to 600 C. for periods of time varying from to 45 minutes. A typical treatment involves cold working a member as the alloy to produce a strain of 10% such as by drawing at temperatures ranging from 25 to 300 C. followed by an anneal at temperatures of from 300 to 600 C. for a period of 30 minutes. The optimum treatment involves cold working to produce a strain of approximately 10% at 25 C. followed by an annealing treatment at 300 C. for 30 minutes.

The results of a series of tests made on bars of zircaloy-4 are set forth in the table below. The physical properties of cold rolled zircaloy-4 bars are 65,000 p.s.i. proof yield and 12% elongation. On a full anneal at 900 C. the yield is 40,000 to 45,000 p.s.i.

actors as structural members and cladding tubes for containing pellets of fissionable material.

It will be understood that the disclosure be construed as illustrative of the invention and not in limitation thereof.

What is claimed is:

1. In a process of producing a zirconium base alloy member having improved tensile strength and ductility properties prior to nuclear radiation, the steps comprising applying a 7% to strain to an alloy member while at a temperature of from about C. to 300 C., the alloy consisting essentially of from 0.1% to 2.5% by weight of tin, and a total of at least 0.1% but not exceeding approximately 2% 'by weight of at least one metal from the Period 4 of the Periodic Table selected TABLE.ROOM TEMPERATgIPRE TEST RESULTS FOR ZIRCALOY-4 ECIMENS Ultimate Elonga- 0.2 proof tensile tion stress, strength, in 2", Treatment Code p.s.i. p.s.i. percent Strain 10% at 25 0.:

As strained A2 57, 817 66, 549 21. 8 Anneal plus mins. at 300 C.;l;50 A4 66, 213 68, 379 37. 2 Anneal plus 30 mins. at 600 C A6 53, 365 66, 507 33. 2 Strain 10% at 300 C.:

As strained B2 61, 860 67, 049 22. 9 Plus 30 mins. at 300 C B4 61, 760 67, 677 24. 2 Plus 30 mins, at 600 C B6 53, 596 65, 695 25.0 Strain 10% at 600 0.:

As strained C2 51, 913 64, 549 34. 3 Plus 30 mins. at 300 C C4 53, 795 62, 040 30.0 Plus 39 mins. at 600 C C0 53, J73 63, 907 28. 0

Proof stress is the stress that will cause a specified small permanent set in the material. The treatment indicated by code A-4 exhibited outstanding room temperature results. Essentially the same results are had at a 15 minute or a 45 minute anneal at 250 C. to 350 C. as with code A-4. An examination of the structures after testing revealed that the materials are not recrystallized by the treatments of A-4 and A-6. Only lattice recovery takes place during the heat treatment after strain. The improved ductility of the zircaloy given either A-4 or A-6 treatment will greatly reduce the probability of brittle failure during service. When irradiated for an equal length of time, where the previously prepared alloy would suffer sufficient lattice damage to be unsafe, the same alloy prepared by the process of the present invention will be completely reliable. In addition to improving ductility, treatment A-4 also increases the allowable design stress for the alloy members.

Accordingly, zirconium base alloys such as zircal0y-2 and zirealoy-4 may be provided with improved values of ultimate strength and ductility to enable them to be better adapted for use as fuel element cladding tubes by preliminarily inducing a nominal stress into the alloy material and subsequently annealing the material at relatively low temperatures for short time periods.

Finally, it is appreciated that novel alloy members are provided having superior tensile strength together with extraordinary ductility which allow members are suitable for greatly extended service time in nuclear refrom the group consisting of iron, nickel, and chromium, carbon not exceeding 0.05% and the balance being zirconium and less than 0.5% by Weight of incidental impurities, and annealing the strained alloy member at a temperature of from 250 C. to 350 C. for a minimum of from about 15 minutes to about minutes, whereby a non-recrystallized member results.

2. The process of claim 1 in which the alloy member is cold worked to induce a strain of 10% at about 25 C. prior to annealing.

3. The process of claim 1 in which the alloy member is cold worked to induce a strain of 10% at 25 C., and in which the alloy member is annealed at a temperature of 300 C. i C. for about thirty minutes.

References Cited WAPD-lZO, Production Annealing of Zircaloy-Z," Goodwin et al., May 25, 1955, pp. 6-12, 14, 15 and 17-20.

BMI-1168, The Mechanical Properties of Zirconium and Zircaloy-2, June 11, 1957, pp. l11.

Johnson: A Study of the Short Time Annealing of Cold Worked Zirconium, Feb. 29, 1956, pp. 1-12.

CHARLES N. LOVELL, Primary Examiner U.S. Cl. X.R. 1481 33 

